MIXER calculates the energy dependent cross section for a composite mixture of up to 10 different materials. Restriction: The present version will calculate only the cross section for one final reaction (ENDF/B section), e.g. total cross section, but not any other reaction. The output from this program has been found extremely useful in the following applications: (1) to calculate a composite total cross section for a later use as a weighting function in self-shielding the cross sections of each constituent of the mixture; (2) to calculate the composite total and fission cross sections in order to calculate the transmission and self-indication through composite materials; (3) to study (via a plotting program) the importance of specific cross section features in the composite cross sections.

IAEA1313/12

This version include the updates up to April 2010.
The 2010 ENDF/B Pre-processing codes process nuclear data formatted in any version of the ENDF formats; ENDF/B-I through ENDF/B-VII evaluations. These codes can be used on virtually any computer: everything from large mainframe computers, to workstations, to IBM-PC (Windows or Linux) and MAC (OSX).
MODIFICATIONS FROM PREVIOUS VERSIONS:
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Mixer VERS. 2010-1 (Apr. 2010): General update based on user feedback.

The user may specify up to 10 different sections of data to be combined. Each section is identified by ZA and MT. If all requested sections are found on the original ENDF/B file, the program will produce a composite section using the union of all energies found in any section. The composite section will not be thinned.

MIXER uses only the ENDF/B BCD format tape and copy all sections except File 3 as read. It is assumed that the data is correctly coded. No error checking is performed. Since File 3 data are in identical format for ENDF/B versions I through VI, the program can be used with all these versions. - The program processes either neutron (MF=3) or photon (MF=23) cross sections, but the two types cannot be mixed. The cross sections to be combined (file 3 or 23) must be in a linearly interpolable form, in ascending energy order of (E, barns). ZA, MF, MT must be in order.

The following programs are all part of the PREPRO2010 package.
ACTIVATE: is designed to create file 10 activation cross sections by
combining file 3 cross sections and file 9 multipliers
COMPLOT: Compares ENDF/B formatted data from two separate input
files. Results are in graphical form.
CONVERT: Automatically converts a FORTRAN program for use on
any one of a variety of: (1) computers; (2) compilers:
(3) precisions; (4) installations; (5) standard or
non-standard file names.
DICTIN: Creates a reaction index for each material.
EVALPLOT:Plots data in the ENDF/B format.
FIXUP: Reads evaluated data in the ENDF/B format; performs
corrections and outputs the results in the ENDF/B format.
GROUPIE: calculates unshielded group averaged cross sections,
Bondarenko self-shielded group averaged cross sections,
and multiband parameters from data in the ENDF/B format.
LEGEND: Calculates linearly interpolable tabulated angular
distributions starting from data in the ENDF/B format.
LINEAR: Converts cross sections in the ENDF/B format (File 3,
23, and 27) to linearly interpolable form (in energy
and cross section) and outputs the result in the ENDF/B
format.
MERGER: Selectively retrieves data by MAT/MF/MT or ZA/MF/MT from
up to 10 ENDF/B data tapes and merges the data into a
single MAT/MF/MT ordered output file.
MIXER: Calculates the energy dependent cross sections for a
composite mixture.
RECENT: Reconstructs energy-dependent cross sections from a
combination of resonance parameters and tabulated
background cross sections in the ENDF/B format.
RELABEL: relabels a ENDF/B preprocessing program so that
statement labels are in increasing order in increments of
10 within each routine.
SIGMA-1: Doppler broadens evaluated cross sections in the
linear-linear interpolation form of the ENDF/B format.
SIXPAK: Checks all double-differential ENDF/B-VI format data (MF=6)
and outputs equivalent uncorrelated data (MF=4, 5, 12, 14, and 15).
SPECTRA: Convert model and general tabulation to linearized spectra (MF=5).
VIRGIN: Calculates uncollided flux and reactions due to transmission of a
monodirectional beam of neutrons through any thickness of material.